Field of the Invention
The present invention generally relates to technologies for reprocessing spent nuclear fuels, and more particularly, to a method for reprocessing a spent nuclear fuel by dissolving a spent nuclear fuel in an aqueous nitric acid solution and separating and recovering nuclides contained in the resulting fuel solution through solvent extraction and also relates to a centrifugal extractor for use in this spent nuclear fuel reprocessing method.
Description of the Related Art
In recent years, use of atomic energy in many countries including Japan has been founded on use of nuclear fuel cycles that include reprocessing of spent nuclear fuels. The reprocessing of spent nuclear fuels is a process of chemically removing FP (fission product) and MA (minor nitride: e.g., Np, Am, Cm, etc.) from the spent nuclear fuels and separating and recovering U and Pu that are reusable in nuclear plants. The reprocessing of spent nuclear fuels not only plays an important part in effective use of energy resources but is also required to comply with demands of nuclear nonproliferation.
Various proposals of spent nuclear fuel reprocessing techniques have been made. A PUREX method, which is one type of solvent extraction method (e.g., refer to Patent Publication 1: Japanese Unexamined Patent Application Publication No. 9-318791), has become a representative reprocessing method due to its excellent selective recovery of U and Pu and nuclear criticality safety. According to a PUREX method, the process proceeds as follows: spent nuclear fuels (fuel assemblies) are chopped into pieces, the pieces are dissolved in nitric acid, the fission products are removed through a solvent extraction process (co-decontamination), the U and Pu are separated from each other through the solvent extraction process, the U is purified and blended with the Pu, the U solution and the U—Pu mixed solution are denitrated, and the U oxides and the U—Pu mixed oxides are produced.
According to existing spent nuclear fuel reprocessing methods, Pu contained in the spent nuclear fuel is ultimately recovered in the form of U—Pu mixed oxides which are then used as MOX fuels for light water reactors. In other words, the methods are designed to prevent ready use of Pu in an isolated form and thus have a certain effect of suppressing nuclear proliferation.
The reprocessing of the spent nuclear fuels through the solvent extraction process has been carried out with pulse column extractors (e.g., refer to Patent Publication 2: Japanese Unexamined Patent Application Publication No. 5-337304), mixer-settler extractors (e.g., refer to Patent Publication 3: Japanese Unexamined Patent Application Publication No. 6-246104), centrifugal extractors (e.g., refer to Patent Publication 4: Japanese Unexamined Patent Application Publication No. 7-108104), etc.
In the above conventional technology, the following matters have been revealed.
(1) The spent nuclear fuel reprocessing requires, from the nuclear management viewpoint, that the U is recovered at a high purity while ensuring that Pu does not exist in an isolated form. Since the existing spent nuclear fuel reprocessing methods include a step of separating U and Pu from each other, these methods are not satisfactory from the standpoint of nonproliferation. Moreover, although the final form of Pu that has been subjected to reprocessing is the U—Pu mixed oxide, highly radioactive FP have already been removed therefrom, thus remaining a risk that Pu could be recovered in an isolated form.
(2) In the devices used for the reprocessing of the spent fuel and including pulse column extractors equipped with partitions (i.e., baffles) inside, in order to increase the efficiency of extracting the nuclides contained in the fuel solutions with a solvent, it is important to increase the number of layers of baffles. Thus, compared to the centrifugal extractors, the pulse column extractors tend to be large in size.
In addition, the phase-separation of the mixed phase of fuel solutions and extraction solvents and the separation and recovery of fuel nuclides are carried out through a settling method by using differences in specific gravity. Thus, the processing rate of solvent extraction is low. This matter is also applied to the mixer-settler extractors which perform the separation of mixed phases and the separation and recovery of fuel nuclides using the settlement method. If the processing rate of the solvent extraction is low, the radiation degradation of the extraction solvents will become severe. Accordingly, in the reprocessing of the spent nuclear fuel with a high radioactive level, such as spent nuclear fuel discharged from a high burn-up reactor and a fast reactor, the lifetime of the extraction solvents tends to be shorter for the mixer-settler extractors in comparison with the use of the centrifugal extractors.